This invention relates to particle fuel (such as is used for high temperature nuclear reactors) of the kind in which the fuel material is coated with a pyrolytic carbon coating impermeable to gaseous fission products and embedded in a matrix of graphite and is concerned more particularly with the testing of such fuel particles to determine the integrity of the coating against any fission gas leakage.
In considering the problem of testing such fuel particles it will be borne in mind that these particles are quite small and that far greater numbers of fuel particles are employed in a high temperature reactor using such carbon coated fuel particles compared with the number of fuel cans in earlier types of graphite moderated gas-cooled reactors operating at lower temperatures such as the Magnox and A.G.R. reactors. There may, for example, be of the order of 10.sup.10 particles in a reactor fuel charge for a high temperature reactor and the level of defect that is required to be measured lies in a range of one in 10.sup.4 to one in 10.sup.6 fully failed particles. It is an object of the present invention to provide a method of testing such particles in a way which is directly relevant to the fission gas release in the reactor and to this standard. One of the requirements is that the testing should be effected before the particles are irradiated in the reactor. With the very large numbers of particles, it is not practicable to load them into the reactor and to check and remove faulty elements after irradiation.
In checking before irradiation, however, a further problem arises with the carbon coated fuel particles in that, although the inner pyrolytic carbon coating has to provide the effective containment of fission gas products, at low temperatures, the matrix graphite will also effectively contain such products although at high temperatures they will diffuse through this matrix graphite.